1. Field of the Invention
The present invention relates to a novel adsorbent for use in a 99Mo/99mTc generator, which is a medical diagnostic, radioisotope generator, and in a 188W/188Re generator, which is a therapeutic radioisotope generator.
2. Description of the Related Art
Technetium-99m (99mTc), which is a very important radioisotope in the medical field, has been used for various types of medical diagnosis. 99mTc, acting as a γ-ray emitter having a half life of 6 hours, is the daughter radionuclide of molybdenum-99 (99Mo), which is produced through neutron absorption or fission of molybdenum-98 (98Mo), and is useful for medical diagnosis of incurable diseases, such as cancers and cardiac diseases.
In addition to the diagnostic use, recently, as an interest is taken on therapeutic radiopharmaceuticals, 188Re is receiving attention as an attractive therapeutic nuclide.
188Re is used for various therapeutic purposes including radioimmunotherapy, synovectomy, and bone pain palliation, and is produced through the decay of the parent nuclide 188W (half life: 69 days), and may be easily obtained in a carrier-free form from a 188W/188Re generator. 188Re has a half life of 16.9 hours and is decayed with β particles, and the β particles have average energy of 764 keV (Emax=2.11 MeV) and emit γ rays (15%) of 155 keV. Therefore, 188Re is advantageous in that an image representing the biodistribution of a compound labeled therewith may be obtained, and pharmacokinetics or uptake quantity in target organs and dosimetry may also be estimated.
Such 99mTc or 188Re may a be repeatedly separated from the parent-daughter mixture using a solvent extraction method or a chromatographic method. The chromatographic method is more easily used than other methods because it needs a small device, is easily operated, and is less limited by temporal constraints. The chromatographic extraction system of 99mTc or 188Re is referred to as a “99Mo/99mTc generator” or “188W/188Re generator”. These generators facilitate the extraction of 99mTc or 188Re in hospitals, thanks to convenience and portability, and thus have been generalized in the field of nuclear medicine throughout the world.
Most 99mTc generators that are presently commercially available utilize 99Mo produced through the fission of highly enriched 235U, and such fission 99Mo has extremely high specific activity, and may thus be adsorbed on a small alumina column (1˜1.5 g of alumina). However, the fission of 235U is disadvantageous because gases and solid radioactive materials are produced in large amounts, undesirably causing waste disposal problems which are burdensome and incur high costs.
Among the many methods for extracting 99Mo from various target materials irradiated in an atomic reactor, U.S. Pat. No. 5,910,971 discloses a method and system for generating 99Mo in the uranyl sulfate nuclear fuel of a homogeneous solution nuclear reactor. In this disclosure, the nuclear fuel containing 99Mo is passed through an organic adsorbent for extracting 99Mo to thus recover 99Mo along a closed-loop path. U.S. Pat. No. 5,962,597 discloses a specific organic adsorbent for extracting 99Mo from the solution nuclear reactor mentioned in U.S. Pat. No. 5,910,971.
Korean Patent Application No. 2002-7007625 discloses an inorganic adsorbent for effectively and selectively extracting 99Mo from an irradiated uranium solution. The adsorbent has high radiation resistance, permitting its use in the high radiation zone of a nuclear reactor. This facilitates a closed cycle extraction process that maintains the uranium concentration of the nuclear fuel through many 99Mo extraction cycles while minimizing radioactive waste disposal problems.
U.S. Pat. No. 4,280,053 discloses a 99mTC generator containing zirconium molybdate (ZrOMoO4) gel produced from 99Mo. In this disclosure, the gel is prepared by dissolving 99Mo in a slight excess of aqueous ammonia or sodium hydroxide solution. Specifically, an acid is added to adjust the pH to 1.5˜7, and the produced solution is added to the stirred aqueous zirconium solution to thus form a molybdate precipitate, which is then collected through filtration or liquid distillation, dried in air, and pulverized to a size suitable for use in the generator, thus obtaining zirconium molybdate.
However, because the method of preparing the zirconium molybdate gel includes a plurality of steps of forming the slurry, adjusting the pH, filtering the slurry, conducting washing and drying, and crushing the resultant precipitate to a preferred particle size, it is technically difficult to produce high-quality radioactive zirconium molybdate gel on a commercial scale through so many steps, and the method is thus undesirable.
Further, 98Mo may be irradiated with neutrons, thereby producing (n,γ) 99Mo. This reaction merely results in 99Mo having low specific activity. In the case where a generator is prepared using the same, the use of a column having a large volume is essential. Accordingly, because the volume of an eluent is also increased, only a 99mTc solution having a low activity concentration is produced.
Currently, the international supply of fission 99Mo is mainly dependant on Canada, and thus there is a need to develop alternatives to technetium generators using fission 99Mo as a feed, in order to realize stable supply and avoid long-distance transport. Because the parent nuclide 118W for generating 188Re is not produced by the fission of highly enriched 235U, the introduction of a generator system using a radioisotope produced by radiating a target material in addition to the fission material onto the nuclear reactor is necessary.
However, in the case where the radioisotope is produced through the irradiation of the target material in addition to the fission material, the production yield, which is in proportion to the square of neutron flux, is greatly decreased under conditions in which the nuclear reactor has low neutron flux. The neutron flux of almost of nuclear reactors in Korea, Europe, and North America is relatively low, within the range from about 2×1014 to about 5×1014 n/cm2·s, so that 99Mo or 188W is obtained in a state having low specific activity. Further, upon the preparation of 188W, 186W needs to be subjected to two neutron capture reactions, leading to further decreased yield. In contrast, diagnostic or therapeutic radioactive atoms require high specific activity, and thus, high-performance adsorbents for radioisotope generators are required.
At present, a generator system using an adsorbent having high adsorption capacity for molybdenum has been devised by Kaken Co., Japan. The generator system utilizes a zirconium polymer (PZC) as an adsorbent and (n,γ) molybdenum (having low specific activity) as a 99mTc source. However, the manufacture of the generator column using the above material suffers in that molybdenum is prepared in a batch type and the solution should be heated for a long period of time for the batch reaction, and thereby complicated radiations are required, and furthermore, even if the column is manufactured, the operating performance thereof is deteriorated.